ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 32

DOE Review Release 05/07/2012

Document Title:

Basis for the Development of an Exposure 

Matrix for Aliquippa Forge, Pennsylvania, 

Period of Operation: January 1, 1947 

through February 28, 1950

Document Number: ORAUT-TKBS-0021

Revision: 01

Effective Date: 04/26/2012

Type of Document: TBD

Supersedes: Revision 00 PC-1

Subject Expert(s): Mutty M. Sharfi

Approval: Signature on File Approval Date: 04/18/2012

Mutty M. Sharfi, Document Owner

Concurrence: Signature on File Concurrence Date: 04/18/2012

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 04/18/2012

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 04/18/2012

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 04/26/2012

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 2 of 32

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

12/21/2004 00 New Technical Basis Document: Basis for the Development of an 

Exposure Matrix for Aliquippa Forge, Aliquippa, Pennsylvania, Period 

of Operation: August 16, 1948 through February 28, 1950. First 

approved issue. Initiated by Cindy W. Bloom.

11/18/2005 00 PC-1 Approved page change revision. Modified to note the covered period 

and the period of operation on page 1, 19 and in Section 2.0 on page 

5. Modified pages 20, 21, and 22 in response to comments and 

guidance from OCAS. Updated Section 1.0 (page 5) standard 

wording. No sections were deleted. Retraining is not required. 

Initiated by Cindy W. Bloom.

Approval:

Signature on File 11/09/2005

Cindy W. Bloom, TBD Team Leader

Signature on File 11/15/2005

Judson L. Kenoyer, Task 3 Manager

Signature on File 11/15/2005

Patricia C. Kimpan, Project Director

Signature on File 11/18/2005

James W. Neton, Associate Director for Science

04/26/2012 01 Revision initiated due to comments from the Advisory Board 

Subcommittee on Dose Reconstruction. Residual period approach in 

Section 5.0 updated per ORAUT-OTIB-0070 and additional data 

capture information. Added an internal approach to account for 

decommissioning and decontamination operations in 1988 and 

1993/1994. Incorporates formal internal and NIOSH review 

comments. Constitutes a total rewrite of the document. Training 

required: As determined by the Objective Manager. Initiated by 

Mutty M. Sharfi.
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 3 of 32

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 5

1.0 Introduction .............................................................................................................................. 7

2.0 Site Description and Operational History.................................................................................. 8

3.0 Estimation of Internal Exposure.............................................................................................. 15

3.1 Uranium Air Sampling................................................................................................. 15

3.2 Occupational Internal Dose Reconstruction Assumptions and Summary.................... 18

4.0 Estimation of External Exposure ............................................................................................ 19

4.1 Submersion and Contamination Exposures ................................................................ 19

4.2 Uranium Billet and Rod Exposures ............................................................................. 20

4.3 Occupational Medical Dose ........................................................................................ 21

4.4 Miscellaneous Information Related to External Dose.................................................. 21

4.5 Occupational External Dose Reconstruction Assumptions and Summary................... 22

5.0 Estimation of Residual Exposure............................................................................................ 22

6.0 Attributions and Annotations .................................................................................................. 24

References ........................................................................................................................................ 25

Glossary ............................................................................................................................................ 30

LIST OF TABLES

TABLE TITLE PAGE

2-1 Documented uranium rollings at Vulcan Crucible ................................................................... 11

2-2 Number of assumed workdays and uranium rolling days ....................................................... 13

2-3 Job titles involved in uranium rolling....................................................................................... 14

3-1 Daily time-weighted average air concentration information..................................................... 16

3-2 Inhalation exposures during rolling operations for Group I workers ........................................ 17

3-3 Inhalation exposures during rolling operations for Group II and III workers ............................ 17

3-4 Annual inhalation exposure during non-AEC operations due to 

resuspension of deposited uranium dust ................................................................................ 18

3-5 Estimated amount of uranium ingested (based on Tables 5 and 7)........................................ 18

3-6 Internal exposure summary for operational period July 23, 1948, to 

February 28, 1950.................................................................................................................. 19

4-1 Calculated photon dose rates for uranium billet and uranium rod........................................... 20

4-2 Direct radiation measurements from September 1948 ........................................................... 20

4-3 External exposure summary for operational period, July 23, 1948 to 

February 28, 1950..................................................................................................................21

5-1 Annual internal and external exposure to residual radioactivity .............................................. 23
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 4 of 32

LIST OF FIGURES

FIGURE TITLE PAGE

2-1 Aliquippa Forge Site, circa 1992...............................................................................................9

2-2 Aliquippa Forge Buildings 3 and 8 where AEC uranium operations occurred ......................... 10
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 5 of 32

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AWE atomic weapons employer

BNI Bechtel National, Inc.

CFR Code of Federal Regulations

cm centimeter

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ft foot

FUSRAP Formerly Utilized Sites Remedial Action Program

GM geometric mean

GSD geometric standard deviation

HEPA high-efficiency particular air

hr hour

ICRP International Commission on Radiological Protection

in. inch

IREP Interactive RadioEpidemiological Program

keV kilovolt-electron, 1,000 electron volts

L liter

lb pound

m meter

MCNP Monte Carlo N-Particle

MDA minimum detectable activity

mg milligram

mL milliliter

mR milliroentgen

mrem millirem

mrep millirep

NIOSH National Institute for Occupational Safety and Health

NYOO New York Operations Office

OCAS Office of Compensation Analysis and Support

ORAU Oak Ridge Associated Universities

PA posterior-anterior

pCi picocuries

POC probability of causation

R roentgen
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 6 of 32

s second

SRDB Ref ID Site Research Database Reference Identification (number)

t ton

U.S.C. United States Code

yr year

μg microgram

μm micrometer

§ section or sections
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 7 of 32

1.0 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) define “performance of duty” for 

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2010).

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 U.S.C. 

7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. While this 

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 

U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally-derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

in dose reconstruction. NIOSH, however, does not consider the following exposures to be 

occupationally derived (NIOSH 2010):

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 8 of 32

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer [AWE] facility” or a “Department of Energy [DOE] facility” as defined in the Energy 

Employees Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 U.S.C. § 7384I(5) 

and (12)]. EEOICPA, as amended, provides for employees who worked at an AWE facility during the 

contract period and/or during the residual period.

Under EEOICPA, employment at an AWE facility is categorized as either (1) during the DOE contract 

period (i.e., when the AWE was processing or producing material that emitted radiation and was used 

in the production of an atomic weapon), or (2) during the residual contamination period (i.e., periods 

that NIOSH has determined there is the potential for significant residual contamination after the period 

in which weapons-related production occurred). For contract period employment, all occupationally 

derived radiation exposures at covered facilities must be included in dose reconstructions. This 

includes radiation exposure related to the Naval Nuclear Propulsion Program and any radiation 

exposure received from the production of commercial radioactive products that were concurrently 

manufactured by the AWE facility during the covered period. NIOSH does not consider the following 

exposures to be occupationally derived (NIOSH 2010):

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

For employment during the residual contamination period, only the radiation exposures defined in 42 

U.S.C. § 7384n(c)(4) [i.e., radiation doses received from DOE-related work] must be included in dose 

reconstructions. Doses from medical X-rays are not reconstructed during the residual contamination 

period (NIOSH 2010). It should be noted that under subparagraph A of 42 U.S.C. § 7384n(c)(4), 

radiation associated with the Naval Nuclear Propulsion Program is specifically excluded from the 

employee’s radiation dose. This exclusion only applies to those AWE employees who worked during 

the residual contamination period. Also, under subparagraph B of 42 U.S.C. § 7384n(c)(4), radiation 

from a source not covered by subparagraph A that is not distinguishable through reliable 

documentation from radiation that is covered by subparagraph A is considered part of the employee’s 

radiation dose. This site profile covers only exposures resulting from nuclear weapons-related work.

Exposures resulting from non-weapons-related work, if applicable, will be covered elsewhere.

2.0 SITE DESCRIPTION AND OPERATIONAL HISTORY

The DOE Office of Worker Advocacy lists the Aliquippa Forge covered period from 1947 to 1950. The 

information that follows supports an assumed period of U.S. Atomic Energy Commission (AEC)

operations at Aliquippa Forge from July 23, 1948, through February 28, 1950, that involved AECcontracted uranium work. This analysis assumed that the residual contamination period extended 

from March 1, 1950, through December 31, 1987, and from January 1, 1989, to December 31, 1992.
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 9 of 32

The Aliquippa Forge radiological source term consisted primarily of natural uranium metal, uranium 

oxides, and natural uranium’s short-lived progeny. Long-lived progeny prevent significant ingrowth 

past 234U in the 238U decay series and beyond 231Th in the 235U decay series. 

Vulcan Crucible Steel Company operated Aliquippa Forge to produce uranium rods for the AEC from 

billets primarily by rolling. Operations with uranium at the forge began when a trial rolling occurred on 

July 23, 1948 (Jones 1948a). The AEC contract for production work (AEC 1948) was initiated on 

August 16, 1948, and was extended through February 28, 1950 (Belmore 1950; Wallo 1981). The 

rolling operation ended on March 30, 1949 (Kelley 1949) with decontamination consuming the rest of 

the contract’s term. Decontamination was completed by Vulcan Crucible in 1950 in accordance with 

then-current AEC guidelines. 

The site consisted of about 19 buildings. The majority of the AEC work occurred in Building 3, the 

rolling mill. Survey results for Building 8 indicate its involvement with uranium activities (Adams and 

Payne 1992a,b). In addition, there were indications of uranium in the locker room, tool room, and 

some areas outside Building 3. Figure 2-1 shows the layout of the site in about 1992, and Figure 2-2

shows the layout of the area that encompassed uranium operations. A sketch from 1948 (Piccot

1948) indicates that at least part of the Building 3/8 area was referred to as C Mill. That sketch 

showed billets in an area known later as Building 8 and a boxcar containing billets and rods to the 

northwest of Building 8; it also showed the lockers, showers, and toilets to the east of Building 3. 

Figure 2-1. Aliquippa Forge Site, circa 1992 (DOE1996a).
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 10 of 32

Figure 2-2. Aliquippa Forge Buildings 3 and 8 where AEC 

uranium operations occurred (DOE 1996a).

The uranium billets that were furnished by the AEC came primarily via boxcar from the Electromet 

facility in New York and the 300 Area on the Hanford Site in Washington (Stroke 1949a). The billets 

were 15 to 28 in. long, 4 to 5 in. in diameter, and weighed from 120 to 270 lb. Vulcan Crucible 

conducted a rolling operation in Building 3 to reduce the billets to rods about 1.5 in. in diameter with 

an increase in length of a factor of 9. The billets were heated and rough-rolled twice to increase the 

length, then finish-rolled and halved before quenching and weighing (Wallo 1981). 

The process at Aliquippa Forge consisted of heating the billets to a temperature between 1,050°F and 

1,100°F. A drag-down operator used a buggy to move the heated billets from the furnace to the north 

side of the roughing roll. The billets were passed through the roughing roll two to four times to 

produce rods of rough dimensions. The rods were then passed through the finishing rolls to achieve 

the desired dimensions. They were then dragged to the shears, cut in two, and dragged back to the 

quenching area for descaling. After the rods were stamped, they were conveyed to the shipping and 

receiving area to be weighed, recorded, and loaded into boxcars (Author unknown ca. 1949; AEC 

1949a; Jones 1948b). 

Little information is available on the actual uranium rollings that took place at Aliquippa Forge during 

the AEC contract period. Documents indicate that each rolling operation involved approximately 
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 11 of 32

96 billets in an 8-hour shift (Stroke 1949b; Reichard 1948). Other documents indicate the Forge had 

a 9-hr workday with a rolling rate of 11 to 12 billets per hour (Breslin 1949), which seems consistent 

with the February 1949 workplace monitoring records, if one additional hour is assigned to lunch and 

breaks. Table 2-1 summarizes the available uranium rolling information. Although some 

documentation indicates that 20% of the time at Vulcan Crucible was spent rolling uranium billets for 

the AEC (Wallo 1981), the AEC contract required that Vulcan Crucible be prepared to perform rolling 

work at least two consecutive weeks out of every five consecutive weeks (AEC 1948), which would be 

40% of the time.

Table 2-1. Documented uranium rollings at Vulcan Crucible.

Date of rolling Type Number of billets rolled

July 23, 1948a Trial 8

August 23–September 2, 1948b Production 822 Type B (80 t)

September 27, 1948c Experimental forging 50

October 21–November 3, 1948d Production 

Experimental forging

Unknown

52 forged

January 3–15, 1949e Production 982 Type B billets

103 Type C billets 

(~120 t)

February 14–23, 1949f Production 112 “small billets per shift”

March 30, 1949g Unknown 70 “large billets per shift”

a. Jones (1948a).

b. Jones (1948b).

c. Schier (1948).

d. Hauff (1948), DOE (2002).

e. Padden (1949).

f. Breslin (1949), DOE (2002).

g. Breslin (1949).

Vulcan Crucible employed about 20 to 25 workers at the rolling plant with approximately 20% of the 

total rolling time occupied by AEC work (Wallo 1981). AEC personnel made occasional visits to 

assess working conditions in the mill and attended some, but not all, rolling operations. 

As of September 2, 1948, the exhaust ventilation system consisted of two large roof ventilators for the 

building (Piccot 1948). The ventilator placement was not considered ideal for the uranium rolling 

operations. 

During the September 1948 AEC visit, a large door and windows at the end of the mill were open, 

which reportedly carried dust to the southeast of the plant (Piccot 1948). The AEC Medical Division 

recommended that Vulcan Crucible upgrade the exhaust ventilation system and install a central 

vacuum to maintain cleanliness. The recommendations were transmitted to Vulcan Crucible on 

October 14, 1948, as requirements for “the contemplated metal rolling contract” (Belmore 1948). A 

November 2, 1948, memorandum indicates that the vacuum system had been installed (Reichard 

1948). However, as of the rolling on February 15 and 16, 1949, the upgrade to the ventilation system 

had not been made (AEC 1949a), although temporary ventilation over the rollers was provided for that 

rolling campaign. 

AEC visits were made before and after modifications to the ventilation system (Belmore 1948; Piccot

1948; AEC 1949a; Breslin 1949; Belmore 1949a). The Monthly Status and Progress Report for April 

1949 (Kelley 1949) states:

Since the Vulcan Crucible Steel Company cannot handle the larger-size 

billets which give better casting yields, it was decided to consolidate 

operation at Simonds. Because of this decision, rolling operations 

ceased at Vulcan after the March [1949] run. Portable Government 
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 12 of 32

property is to be removed immediately. Attached property is to be left at 

Vulcan, provision will be made for emergency standby facilities until the 

contract expires in August. 

During World War II, permissible levels for uranium dust in air were set at 500 μg/m3 for insoluble 

uranium compounds and 150 μg/m3 for soluble uranium compounds. After the war, the University of 

Rochester lowered its recommendation for soluble uranium compounds to 50 μg/m3 based on 

chemical toxicity, which for natural uranium is equivalent to 70 dpm/m3

. This level was based 

primarily on animal studies. The Medical Division of the AEC New York Operations Office (NYOO) 

felt that a "maximum permissible level" was unknown and should be based on human data. 

Therefore, the 50-μg/m3 level was referred to as the "preferred level" (AEC 1949b). 

The AEC visited Aliquippa Forge on September 1, 1948, to conduct a health physics survey and to 

determine (Piccot 1948):

1. The type of physical examination given to the men.

2. The use of protective clothing, showers, and lunchroom.

3. The exposure of personnel to direct radiation.

4. The spread of contamination in the plant.

5. The concentration of radioactive dust in the air.

Medical examination requirements for Aliquippa Forge uranium workers were specified in October 

1948 (Belmore 1948). The AEC staff noted on March 29, 1949, that “a complete blood count, a urine 

and a chest X-ray were done on all employees exposed in the rolling area before work on the AEC 

Project was started” (Tabershaw 1949) and “a repeat blood count and urine [medical, not radioactivity 

analyses] were done on all workers after a period of 6 months.” The X-ray examinations were 

performed by Dr. W. T. Rice, who kept the films in his Rochester, Pennsylvania, office.

As of the September 1, 1948 visit, there were few health physics controls in place. Piccot (1948) 

noted, “workers furnish and launder their own clothes, shoes and gloves and usually change before 

going home. Separate lockers are not provided for clean and dirty clothes. Smoking is permitted in 

the mill and some men were observed smoking with their dirty gloves on. The men also eat their 

lunch in the vicinity of the mill.” In addition, the report noted that workers were loitering or standing 

near uranium billets or rods, and one even sat on billets while taking a break. 

In October 1948, the AEC specified “protective clothing and hygienic procedures” (Belmore 1948). 

Protective clothing was listed as dedicated work clothing, gloves, and shoes. The AEC recommended 

that Vulcan Crucible provide workers with clothing that could be kept separate from personal clothes. 

Workers were to be instructed to not eat or smoke with gloved hands, and to wash hands thoroughly 

before smoking, eating, or leaving the shift. The AEC also recommended showering at the end of the 

shift. No later reports were found to indicate whether the company supplied clothing, but there was 

indication that some workers showered and changed at the end of the shift. The available records 

indicate contamination controls were not strictly implemented. 

Air samples were taken during the September 1948 visit; one sample (during the third pass in back of 

the mill) showed an air concentration as high as 1,800 times the preferred level (i.e., 50 μg/m3

). 

Although peak values need to be considered, the fact that work tasks and worker locations were 

constantly changing resulted in time-weighted exposures that were typically much lower than the peak 

values. The other air sampling results for September 1 and 2, 1948, showed concentrations in the 

range of 2.6 to 510 times the preferred level, with 15 of the 22 results less than or equal to 50 times 

the preferred level. 
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 13 of 32

A NYOO report of an AEC visit to the Aliquippa Forge on February 15 and 16, 1949, described timeweighted radioactive dust exposures between 2.7 and 5,300 times the preferred level depending on 

the type of job (AEC 1949a). A review of the February 1949 report and calculations showed that 

February report was in error and overstated the maximum exposure by about a factor of 10. When 

discussing the higher concentration, the report stated: 

It was noted during the sampling that relatively large flakes of scale were being thrown 

from the rods at this operation. The above concentrations may therefore represent 

some number of large, non-respirable particles, and not be a true indication of 

exposure. 

A later summary report (AEC 1949b) stated that four workers who were directly involved in the rolling 

were exposed to as much as 530 times the preferred level. The summary report mentioned that the 

recommended ventilation system upgrade had not been installed, but that it was on order and that a 

temporary although inadequate system was in place. 

The AEC visit of March 30, 1949, occurred after the installation of a new exhaust fan with twice the 

capacity of the previous fan. During this visit, the largest calculated time-weighted uranium exposure 

was 21 times the preferred level. A reevaluation of one worker’s exposure based on a statement that 

one sample appeared to be unreasonably low indicates that the largest exposure might have been 

57 times the preferred level. This is still indicative of a reduction of uranium air concentrations due to 

the improved ventilation. AEC noted during this visit that rather than rolling 112 “small” billets during a 

9-hour shift, 70 “larger” billets were rolled in an 8-hour shift, which might also have contributed to the 

lower uranium dust concentrations (Breslin 1949). The AEC record indicates that there were no 

rolling operations after March 1949 and that only cleanup operations were taking place (Kelley 1949).

The purpose of the April 24, 1949, AEC visit was to survey the contamination from previous rollings. 

The visit resulted in recommendations for the cleanup of the mill. The visit of July 21, 1949, evaluated 

the effectiveness of the recommended decontamination. During these two visits (Belmore 1949a,b), 

AEC took direct measurements of surface contamination using a Zeuto, but collected no smear 

samples. (A Zeuto is a portable ionization chamber. The early models were used to measure alpha 

contamination; some models also measured beta and gamma radiation.)

Based on the time-motion information AEC collected on February 15 and 16, 1949 (AEC 1949a) and 

on March 30, 1949 (Breslin 1949), this evaluation assumed 10-hour workdays before March 30, 1949, 

and 8-hour workdays thereafter. Table 2-2 lists the assumed number of workdays and uranium rolling 

days in each period.

Table 2-2. Number of assumed workdays and uranium rolling days. 

Start End

Nonuranium 

rolling workdays Rolling days Calendar days

07/23/1948 12/31/1948 116 50 162

01/01/1949 03/29/1949 63 30 88

03/30/1949 12/31/1949 198 90 277

01/01/1950 02/28/1950 42 20 58

The analyses for this exposure matrix divided the workers into three categories as listed in Table 2-3. 

While different tasks in the mill resulted in differences in exposures (AEC 1949a), it is not known if 

each worker always performed the same task within a group, or if workers temporarily worked in 

locations where higher or lower exposures occurred. Workgroup exposure assignments are based on 

data that are suggestive of worker exposures and further modified by uncertainty parameters to 

ensure that the reconstructed dose distributions capture the larger exposures. Depending on the 
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 14 of 32

organ of interest and the ancillary data on a specific claim, additional considerations might be 

appropriate. 

Table 2-3. Job titles involved in uranium rolling.

Group I Group II Group III

Hook Man (Front Hooker, Back Hooker) Shipping and Receiving Guard

Drag-Down Operator (Buggy Man) Checker Technical Supervisor

Rougher Rod Weigher Office Workers

Finisher

Shear Man

Shear Man Helper

Heater (Furnace Man)

Heater Helper (Furnace Man Helper)

Mill Roller

Catcher

Hot Sawyer 

Hot Sawyer Helper

Rod Stamper

Quencher

The analyses assumed:

• Group I workers were involved primarily with heating, rolling, finishing, and quenching of 

uranium. They spent the largest part of their time in the general mill area near the furnace and 

rollers.

• Group II workers worked primarily with the finished uranium rods and spent the largest part of 

their time in the rod storage area and boxcar. 

• Group III workers were not engaged directly in the processing of uranium. They probably had 

lower internal and external exposures than the other groups, with the possible exception of the 

guards, who might have spent time near the boxcar and the uranium storage areas. 

Table 2-3 categorizes Aliquippa Forge uranium rolling job titles in the designated groups. Exposures 

to workers who do not fit in the groups are designated to an “unknown” category, which is included in 

the internal and external exposure sections of this document. 

In July of 1949, a survey was performed to determine forge cleanup requirements (Belmore 1949c). 

Additional AEC assessments and surveys were made throughout the cleanup process. Although AEC 

noted that the cleanup personnel had no monitoring equipment, the assessments concluded that a 

sufficient job of cleanup had been done (Eisenbud 1950; Belmore 1950). 

The Formerly Utilized Sites Remedial Action Program (FUSRAP) began in 1976. Perry (1993) stated:

A radiological survey in 1978 identified contamination (primarily uranium-238) in and 

around onsite buildings. The site was designated for further remediation under 

FUSRAP, and the small operation was shut down and the building evacuated. 

Because 238U is the predominant isotope by mass in natural uranium and is more easily identified than 

the other isotopes, the 1978 survey report might have been referring to natural uranium, which 

consists of approximately equal activities of 234U and 238U and a smaller amount of 235U. Reported 

238U quantities might include all of the uranium activity or just part, depending on actual analysis 

techniques and reporting procedures. 
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 15 of 32

Radioactive contamination was found during the survey of May 2 to 8, 1978, on the dirt floor, concrete 

floor, steel floor plates, and the overhead beams above the furnaces that were used in the uranium 

processing (Wynveen et al. 1982). The actual date of building evacuation is unknown and is assumed 

to have occurred on May 31, 1978, after the FUSRAP survey.

In August 1983, the Aliquippa Forge site was designated for remedial action under FUSRAP (DOE 

1996a). In December 1987, storage activities began in Building 3. Interim remedial actions were 

taken from October to December 1988 to enable additional restricted use of Building 3 for expansion 

of a small forging operation (Baublitz 1988; Harbert 1989; DOE 1996a). Controlled areas were 

established to prevent access to contamination (Seay 1988; DOE 1996a). The exact date of 

reoccupation is unknown, but is assumed to have occurred as early as December 1, 1987. As of May 

17, 1993, the buildings were no longer in use, although the actual last date of use is not clear. This 

analysis assumed that a second period of residual exposure occurred from December 1, 1987, to May 

17, 1993. Final remedial activities occurred from about June 1993 to September 1994 (Abelquist 

1995; DOE 1996a). A final survey and decontamination were performed in 1995 (DOE 1996a). A 

DOE notice of certification was published in October 1996 in the Federal Register (61 FR 55981).

Because recycled uranium was not available to contracted AEC facilities until after March 1952 (DOE 

2001), this exposure analysis did not consider it. 

3.0 ESTIMATION OF INTERNAL EXPOSURE

The primary source of internal radiation exposure at Aliquippa Forge was uranium dust from the 

manipulation and oxidation of uranium metal during the rolling and related processes. There is no 

indication that uranium rolled at Vulcan Crucible was enriched, so the analysis assumed natural 

uranium enrichment. 

Human and animal studies have indicated that oxides of uranium can be very insoluble (ICRP 1995), 

indicating absorption type S. Other in vitro dissolution studies of compounds at uranium facilities have 

shown that oxides of uranium exhibit moderate solubility (Eidson 1994; Heffernan et al. 2001) 

suggesting absorption type M. In vitro dissolution tests on oxides produced from uranium metal 

during depleted uranium armor penetrator tests have indicated multicomponent dissolution rates, with 

25% of uranium dissolving with a half-time of less than or equal to 0.14 days and 75% dissolving with 

a half-time of 180 days. Because there was no specific information on the solubility of aerosols at the 

forge, this analysis assumed that both types M and S were available. The selection of absorption type 

should depend on the organ of interest. Dose reconstructions should assume International 

Commission on Radiological Protection (ICRP) Publication 66 default parameters for particle

deposition (ICRP 1994).

The uranium fusion photofluorimetry urinalyses performed by the University of Rochester and the 

AEC NYOO were similar to those performed at the Fernald Plant. The default detection threshold for 

uranium urinalysis is 14 μg/L, based on the Fernald reevaluation of its uranium fluorimetry urinalysis 

minimum detectable activity (MDA) in 1993 (Blalock 1993).

Individual uranium urinalysis data are available for some forge workers (AEC 1949c). For 

unmonitored workers or unmonitored periods, this Technical Basis Document analyzes air monitoring 

data for use in reconstructing internal doses. 

3.1 URANIUM AIR SAMPLING

Air sampling was performed at Aliquippa Forge during some of the uranium rolling (AEC 1949b; 

Breslin 1949). The air samples consisted of collection on filters of radioactive particulate from 

breathing zones, general areas, processes, and effluents. The measured alpha activity on the filter 
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 16 of 32

was used to determine airborne alpha activity concentrations. The AEC matched these air 

concentration determinations with information about worker categories, locations, tasks, and time at 

each location or task. For some tasks and locations, multiple samples were collected; the mean count 

rate was calculated and used to calculate an average air concentration. 

The AEC used the information on work tasks with the measured air concentration to determine an 

average air concentration weighted by time and summed these weighted average air concentrations 

to determine a daily weighted average air concentration for specified job categories. These air 

concentration results are further analyzed here to determine group geometric means (GMs). The 

daily weighted average air concentrations were weighted further by the AEC-reported number of 

workers who were exposed at a given concentration. The GMs of the workgroups’ daily weighted 

average air concentrations were calculated. The geometric standard deviations (GSDs) of both the 

job category concentrations and the workgroup concentrations were determined to provide an 

indication of the distribution of the data. The latter GSD includes consideration of the number of 

people in each category in the AEC study. This analysis assumed that, because data are limited and 

rigorous analyses to determine distribution type are not likely to be meaningful, a lognormal 

distribution could represent the time-weighted exposures and the subsequently derived annual organ 

doses. 

The report on the February 1949 rolling shows time-weighted air concentrations (based on 

measurements at the plant) before the upgrade of the ventilation system. The report on the March 

1949 rolling (Breslin 1949) shows the time-weighted air concentrations after the upgrade of the 

ventilation system. During spot checks of the February 1949 air concentrations, this analysis revealed 

a mathematical error for the Hook Man’s daily weighted air concentration, which resulted in a change 

from the 36,800 dpm/m3 to 38,200 dpm/m3

. An error was also found for the February result for 

Roughers, but the change was small and correction of the error would have slightly reduced the 

exposure, so it was ignored. The March report noted that “one unsubstantiated sample of unusually 

low concentration” probably resulted in an “erroneous” daily weighted average air concentration for 

the Hook Man. To remedy this, the March 30 average ratio of the second-to-first rolling pass air 

concentrations was determined. This ratio, 19.2 to 1, was used to modify the low air concentration 

result. The two sets of data from February and March 1949 were then used to estimate the air 

concentrations to determine internal exposures by workgroup. Table 3-1 lists the GMs and GSDs for 

the job category and the workgroups’ daily time-weighted average air concentrations.

Table 3-1. Daily time-weighted average air concentration information.

Air samples collection date

Group I Group II and III

02/15/1949 03/30/1949 02/15/1949

Job categories Number of categories 9 9 6

GM (dpm/m3) 2,210 545 484

GSD 5 3.2 2

Workgroup Number of workers 15 9 10

GM (dpm/m3) 2,610 479 608

GSD 5.3 3 1.8

To be favorable to claimants, the analysis of intakes based on air concentrations assumed that 

uranium rolling took place between July 23, 1948, and February 28, 1950. It also assumed that 10 

days of every month were spent rolling uranium because the AEC required Vulcan Crucible to be 

prepared to spend 2 consecutive weeks of every 5 consecutive weeks performing AEC work (AEC 

1948). Rolling did not occur after March 1949, and data to estimate exposure directly from cleanup 

operations were not available. The analysis assumed that the internal exposure rates during cleanup 

would not have exceeded the internal exposure rates during rolling operations with the improved 

ventilation. 
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 17 of 32

The breathing rate is based on the default for light work in ICRP Publication 66 (ICRP 1994, Table 6, 

p. 23). The intakes, in picocuries, were calculated by dividing the GM of a workgroup’s time-weighted 

air concentration by 2.22 dpm/pCi and multiplying this result by the breathing rate and the assumed 

number of exposed hours at the given concentration. Aliquippa Forge internal organ doses were 

assumed to be lognormally distributed and, the GSDs for the calculated internal organ doses were 

assumed to be 5.3. The primary bases for selecting a GSD of 5.3 for all internal organ dose 

calculations was to simplify and expedite dose reconstructions and to encompass the largest 

distribution from the air sampling data. Several assumptions included in the intake and dose 

reconstruction are likely to be overestimating assumptions, which increases the estimate of the 

median intakes. This overestimation of the median, combined with the assumed GSD of 5.3, is 

believed to be sufficiently large to describe the organ dose distributions.

Tables 3-2 and 3-3 list estimated annual inhalation intakes during rolling that were assigned to 

workers in each category. Because air concentrations were not available for Groups II and III in 

March 1949, the March-to-February ratio of the Group I GM air concentration (4 to 1), was used to 

estimate a Group II and III air concentration for March 30, 1949.

Table 3-2. Inhalation exposures during rolling operations for Group I workers.

Work period

Number 

of 

months

Number of 

potential AEC 

workdays

Air 

concentration 

(pCi/m3)

Breathing 

rate (m3/hr)

Hours 

worked 

per day

Intake 

(pCi)

07/23/1948–03/29/1949 8 80 1,180 1.2 10 1.13E+6

03/30/1949–02/28/1950 11 110 216 1.2 8 2.28E+5

Total 1.36E+6

Table 3-3. Inhalation exposures during rolling operations for Group II and III workers.

Work period

Number 

of 

months

Number of 

potential AEC 

workdays

Air 

concentration 

(pCi/m3)

Breathing 

rate (m3/hr)

Hours 

worked 

per day

Intake 

(pCi)

07/23/1948–03/29/1949 8 80 274 1.2 10 2.63E+5

03/30/1949–02/28/1950 11 110 68.5 1.2 8 7.23E+4

Total 3.35E+5

There was a potential for internal exposure to resuspended material from the AEC work during 

non-AEC operations. To estimate exposure from resuspended materials, this analysis assumed that 

surfaces in the building became contaminated by deposition of uranium dust during rolling operations. 

The level of contamination was determined by multiplying the largest air concentrations, listed in 

Table 3-2, by the indoor deposition velocity and the assumed deposition time. Indoor deposition 

velocity is dependent on the physical properties of the room (such as air viscosity and density, 

turbulence, thermal gradients, and surface geometry). It is also dependent on the physical properties 

of the aerosol particles (such as diameter, shape, and density). In this case, these characteristics are 

not known, so the terminal settling velocity was calculated for an aerosol with the ICRP Publication 66 

default particle size distribution of 5-μm activity median aerodynamic diameter (ICRP 1994). The 

calculated terminal settling velocity was 7.5 × 10-4 m/s, which is within the range of measured 

deposition velocities (2.7 × 10-6 to 2.7 × 10-3 m/s) in various studies (Biwer et al. 2002).

The calculated surface contamination level from airborne dusts during the uranium rolling from July 

23, 1948, to February 28, 1950, was 5.35 × 106 pCi/m2 (119,000 dpm/100 cm2

). To be favorable to 

claimants, the analysis assumed that all of the surface contamination was present for the entire period 

of AEC operations. Using a resuspension factor of 1 × 10-6

/m (Abu-Eid et al. 2002), the air 

concentration due to resuspension would have been 5.35 pCi/m3

. Table 3-4 lists the assumed annual 

inhalation intake from resuspension of deposited material. The intakes in Table 3-4 are added to the 

intakes in Table 3-2 or Table 3-3 before calculation of annual organ dose (see Table 3-6).
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 18 of 32

Table 3-4. Annual inhalation exposure during non-AEC operations due to resuspension of deposited 

uranium dust.

Work period

Hours 

per day 

Workdays per 

work period 

Breathing 

rate (m3/hr)

Resuspended air 

concentration (pCi/m3) Intake (pCi)

07/23/1948–03/29/1949 10 179 1.2 5.35 1.15E+4

03/30/1949–02/28/1950 8 240 1.2 5.35 1.23E+4

Total 2.38E+4

When inhalation intakes are calculated from air concentrations, ingestion intakes must also be 

considered. NIOSH (2004) states that the daily ingestion rate in picocuries can be estimated by 

multiplying the daily air concentration in picocuries per cubic meter by a factor of 0.2 for an 8-hour 

workday. For a 10-hour workday, the multiplier would be 0.223. The daily ingestion rates during AEC 

uranium work were estimated from the air concentrations in Table 3-2, which are larger than the 

concentrations in Table 3-3. The daily ingestion intakes from resuspended uranium were estimated 

from Table 3-4. The ingestion intake rates were multiplied by the number of exposed workdays at the 

calculated levels and summed. The ingestion intakes in Table 3-5 apply to workers in Groups I, II, 

and III. 

Table 3-5. Estimated amount of uranium ingested (pCi) (based on Tables 5 and 7).

Work period

Days of 

uranium 

rolling

Uranium ingestion 

rate (during uranium 

rolling) (pCi/d)

Non-uranium 

rolling 

workdays

Uranium ingestion 

rate (during normal 

operation) (pCi/d)

Intake 

(pCi)

07/23/1948–03/29/1949 80 2.62E+02 99 1.19 2.12E+4

03/30/1949–02/28/1950 110 4.32E+01 130 1.07 4.89E+3

Total 2.62E+4

3.2 OCCUPATIONAL INTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

The assumed uranium photofluorimetry urinalysis MDA is 14 μg/L. 

The assumed operational exposure period was from July 23, 1948, to February 23, 1950. The 

uranium-rolling period was over by March 31, 1949, and the analysis assumed that the exposure after 

that time was due to cleanup activities, which might have resulted in changing uranium intake rates for 

some workers. If limited bioassay data are used to calculate intakes, the assignment of the exposure 

period needs to be considered carefully, because assuming that bioassay results are collected during 

periods of elevated exposure, when the result is actually collected days or months after the period of 

elevated exposure, could result in underestimation of intake. For individuals unlikely to have been 

involved in cleanup, but who have positive bioassay results, it is reasonable to set the intake period to 

the period of uranium rolling operations (July 23, 1948, through March 30, 1949) and to add an 

additional intake based on the Table 3-6 intakes for the period after the last bioassay. 

For unmonitored workers or unmonitored periods, Table 3-6 lists intake rate assumptions for 

natural uranium. The intake mode is chronic. The dose distribution is assumed to be lognormal with 

a GSD of 5.3. 
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 19 of 32

Table 3-6. Internal exposure summary for operational period July 23, 1948, to 

February 28, 1950.

Start End

Intake 

route

Absorption 

type

Intake 

(pCi/day)

Group I and unknown 07/23/1948 03/29/1949 Inhalation M, S 4,610

07/23/1948 03/29/1949 Ingestion (a) 85.1

03/30/1949 02/28/1950 Inhalation M,S 731

03/30/1949 02/28/1950 Ingestion (a) 14.6

Group II and III 07/23/1948 03/29/1949 Inhalation M, S 1,120

07/23/1948 03/29/1949 Ingestion (a) 85.0

03/30/1949 02/28/1950 Inhalation M,S 267

03/30/1949 02/28/1950 Ingestion (a) 14.6

a. Choose same f1-value as used for inhalation per NIOSH (2004).

4.0 ESTIMATION OF EXTERNAL EXPOSURE

Individual external dosimetry results for Aliquippa Forge workers are not available. If results are 

found, dose reconstructors should consider those results in the evaluation of external dose.

For dose reconstruction, this analysis assumed there was a potential for external exposure to natural 

uranium metal from five sources: 

• Submersion in air contaminated with uranium dust,

• Exposure from contaminated surfaces,

• Exposure to electrons from the surface of the uranium billets and rods,

• Exposure to photons from the uranium billets and rods, and

• Exposure to an annual diagnostic X-ray.

The majority of photons from natural uranium metals have energies in the range of 30 to 250 keV. 

Solid uranium objects provide considerable shielding of the lower energy photons and harden the 

spectrum, which causes the majority of photons from a solid uranium object, such as a billet or a rod, 

to have energies greater than 250 keV. While solid uranium sources have a hardened photon 

spectrum, exposure to a thin layer of uranium on a surface results in a larger fraction of exposure to 

lower energy photons. To be favorable to claimants, the analysis assumed workers were exposed to 

photon energies in the range of 30 to 250 keV. Nonpenetrating dose from natural uranium consists 

primarily of electrons with energies above 15 keV. For consistent presentation, exposure or dose is 

reported as either penetrating or nonpenetrating and is assumed to be associated with photons with 

energies of 30 keV or greater, and with photons with energies less than 30 keV or with electrons, 

respectively. 

4.1 SUBMERSION AND CONTAMINATION EXPOSURES

In a survey at Simonds Saw and Steel, the AEC suspended 20 film badges about 5 ft from the floor in 

the rolling mill for 192 hours “to determine the long term direct [external] radiation to individuals” (AEC

1949d). When the badges were retrieved, they were covered with radioactive dust from the plant, 

which probably resulted in an overestimate of the true area radiation levels. The maximum results 

were reported as 5.6 mR/hr beta and 0.34 mR/hr gamma. The results of these measurements were 

assumed to be representative of the general levels of external exposure from submersion in air and 

contaminated surfaces at Aliquippa Forge. This analysis assumed that the data distribution was 

lognormal. The calculated GMs were 1.3 mR/hr with a GSD of 2.3 for the nonpenetrating radiation 

and 0.26 mR/hr with a GSD of 1.2 for the penetrating radiation. This assumption does not appear to 

be inconsistent with the reported Zeuto (portable ionization chamber) beta and gamma readings at 

Simonds Saw and Steel of 2 mR/hr or less for most areas (AEC 1949d), some of which appear to be 

contact readings. The beta reading was assumed to be related to the nonpenetrating dose, and the 
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 20 of 32

gamma reading was assumed to be related to the penetrating dose. These assumed exposures 

during operational years are listed in Table 4-3. The analysis assumed all workers were exposed to 

penetrating and nonpenetrating radiation from submersion in air and contamination for each workday 

for 10 hr/d before March 30, 1949, and 8 hr/d thereafter. 

4.2 URANIUM BILLET AND ROD EXPOSURES

Another assumption was that workers received a deep dose from photon exposure to the uranium 

billets and rods. According to reports, the AEC work involved rolling uranium billets 4 to 5 in. in 

diameter and 15 to 28 in. long. The billets were rolled into rods approximately 1.5 in. in diameter and 

20 ft long. Monte Carlo N-Particle (MCNP) calculations determined the photon (including 

bremsstrahlung) dose rate at the surface, 1 ft, and 1 m from a 5-in.-diameter by 28-in.-long cylindrical 

billet and a 1.405-in.-diameter by 20-ft-long rod. Table 4-1 lists calculated photon dose rates for the 

uranium billet and rod.

Table 4-1. Calculated photon dose rates for uranium 

billet and uranium rod.

Distance from source

Billet dose rate

(mrem/hr)

Rod dose rate

(mrem/hr)

Surface 7.74 5.09

1 ft 0.703 0.285

1 m 0.108 0.0883

The analysis assumed Group I workers were exposed primarily to the billet dose rate and that 

Group II workers were exposed primarily to the rod dose rate. It also assumed that the dose rate at 

1 ft was the median dose rate and the dose rate at the surface was the upper 95th percentile. 

The annual penetrating dose rates in Table 4-3 were calculated by multiplying the median photon 

dose rates by the number of rolling days per year and the number of work hours (10 hr/d before 

March 30, 1949, and 8 hr/d thereafter).

Shallow doses from the uranium billets and rods were estimated using the measurements in 

Table 4-2. These measurements were taken during an AEC survey in September 1948 (Belmore 

1948).

Table 4-2. Direct radiation measurements from September 1948.

a

Location of measurement

Dose rate

(mrep/hr)

b

GROUP I

Contact with floor next to the quench tank where oxide scale collected 8

Contact with floor in front of rolls where oxide scale collected 5–10

Same location but 18 in. high 2–5

GROUP II

4 ft above a pile of rods in the boxcar 20

5 ft from the end of a pile of rods next to the door of the boxcar 5

2 ft from the end of the same pile 13

a. Belmore (1948).

b. A rep is an obsolete unit of dose equivalence (roentgen-equivalent-physical) approximately equal 

to a rem.

This analysis estimated the shallow dose for Group I by assuming that the median dose rate was 

5 mrem/hr and that the upper 95th-percentile dose rate was 10 mrem/hr, giving a GSD of 1.5. For 

Group II, the assumed median dose rate was 5 mrem/hr and the assumed upper 95th-percentile dose 

rate was 20 mrem/hr, giving a GSD of 2.3. These exposure rates were multiplied by the assumed 

number of uranium rolling hours in the period. Table 4-3 lists these doses for Group I and II workers. 
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 21 of 32

The analysis assumed that Group III workers were unlikely to be in close contact with the rods and 

billets for extended periods; it also assumed that air and surface external exposures account for the 

external exposures. 

Table 4-3. External exposure summary for operational period, July 23, 1948 to February 28, 1950.

Workers

Exposure 

mode

Exposure 

type

Exposure or 

dose rate

Basis/ 

Photon DCF

Exposure 

time 

assumption Year

Annual

exposure

IREP 

distribution

All

Submersion/ 

area 

contamination

Penetrating 0.26 mR/hr Film badge / 

Exposure

2000

work-hr/yr

1948

1949

1950

0.302 R

0.576 R

0.087 R

Lognormal

GSD 1.2

Nonpenetrating 1.3 mR/hr Film badge / 

Exposure

2000

work-hr/yr

1948

1949

1950

1.508 R

2.878 R

0.437 R

Lognormal

GSD 2.3

Medical X-ray n/a n/a ORAUT 

2011b n/a n/a n/a

Group I U billets

Penetrating 0.703 

mrem/hr

MCNP 

calculation / 

Dose 

Equivalent 

(Hp(10))

3 hr/rollingday

1948

1949

1950

0.105 rem

0.253 rem

0.042 rem

Lognormal

GSD 4.2

Nonpenetrating 5 mrep/hr

Instrument 

measurement

/ NA

3

hr/rollingday

1948

1949

1950

0.750 rem 

1.800 rem 

0.300 rem

Lognormal

GSD 2.7

Group II 

and 

unknown

U rods

Penetrating 0.285 

mrem/hr

MCNP 

calculation / 

Dose 

Equivalent 

(Hp(10))

7

hr/rollingday

1948

1949

1950

0.100 rem 

0.239 rem 

0.040 rem

Lognormal 

GSD 5.7

Nonpenetrating 5 mrem/hr

Instrument 

measurement

/ NA

7

hr/rollingday

1948

1949

1950

1.750 rem

4.200 rem 

0.700 rem

Lognormal 

GSD 2.3

4.3 OCCUPATIONAL MEDICAL DOSE

According to a March 29, 1949 memo (Tabershaw 1949), X-ray procedures were taken off-site at a

physician’s office in Rochester, PA. Therefore, per guidance in Guidance on Assigning Occupational 

X-Ray Dose Under EEOICPA for X-Rays Administered Off Site (ORAUT 2011b), no occupational 

medical dose should be assigned.

4.4 MISCELLANEOUS INFORMATION RELATED TO EXTERNAL DOSE

This section includes external dose information that might be of interest for specific dose 

reconstructions, but that this analysis did not consider generically because of its limited applicability or 

because of limited information.

In September 1948, Piccot (1948) noted that some workers were smoking while wearing dirty gloves 

and one worker was observed sitting on billets during a break. These activities are not directly 

considered in the external dose evaluations. They were observed early in the Vulcan Crucible 

uranium rolling operations and might have been limited occurrences. 

During the 1949 decontamination activities, an AEC inspector noted (Belmore 1949a):

There were 15 pairs of shoes in two steel drums in the storeroom of the plant. The 

soles of the shoes gave an average reading of 14,000 alpha d/m and 1.3 mr/hr betagamma. The leather tops of the shoes showed an average reading of 5,000 alpha d/m 

and less than [0.4] mr/hr beta-gamma. Inside the shoes there were negligible alpha 

and beta-gamma readings.
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 22 of 32

4.5 OCCUPATIONAL EXTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

Table 4-3 summarizes occupational external doses during uranium operations at Aliquippa Forge.

5.0 ESTIMATION OF RESIDUAL EXPOSURE

After the end of AEC rolling operations, a July 1949 survey was performed. The survey indicated that 

the maximum air dust concentration, taken during normal operations in the Furnace area, was 

5.9 μg/m3 or 8.94 dpm/m3 (assuming a specific activity of 1.516 dpm/μg for natural uranium) (Belmore 

1949b). 

Vulcan Crucible received essentially pure uranium metal (no radium) for processing. This is 

confirmed by the 1978 Argonne National Laboratory survey (Wynveen et al. 1982), which showed 

radon levels varying between 0.11 and 0.27 pCi/L (0.0.0011 to 0.0027 Working Level assuming 100% 

equilibrium), which are within the normal range of atmospheric radon content of 0.1 to 0.5 pCi/L 

(Eisenbud 1987, p. 417). The building was reportedly evacuated in 1978 (Perry 1993); this is 

assumed to have occurred on May 31, 1978, after the Argonne National Laboratory survey.

The Aliquippa Forge site was included in the DOE FUSRAP in August 1983. In December 1987, 

Bechtel National, Inc. (BNI) surveyed Aliquippa Forge for the purpose of allowing the use of portions 

of Building 3 for storage. Interim remedial activities were conducted by BNI in 1988 by removing 

contaminated materials and equipment and placing a barricade around the remaining contaminated 

area (Adams and Payne 1992b, p. 12). DOE noted that access to the contaminated areas was not 

allowed (Seay 1988). 

In 1992 and 1993, areas in and adjacent to Buildings 3 and 8 were further characterized (Abelquist 

1994; Adams and Payne 1992a,b). The maximum reported exposure rate at 1 m was 0.014 mR/hr

(Adams and Payne 1992a). The maximum removable surface contamination was 350 dpm 

alpha/100 cm2

. 

To calculate internal exposure from residual activity the analysis assumed that all buildings had an air 

concentration of 8.94 dpm/m3 in 1950. This operational air concentration was assumed to have 

occurred for 1 year with no cleanup. An indoor deposition velocity of 0.00075 m/s was applied to 

calculate a 2.11 × 105 dpm/m2 surface contamination level at the end of operations/start of the 

residual period. A resuspension factor of 1 × 10-6 m-1 was applied to the surface contamination level, 

resulting in an air concentration of 0.211 dpm/m3

. A source term depletion rate was calculated based 

on a starting air concentration in 1950 and the air concentration calculated based on the 1992 survey 

data (ORAUT 2012). The 1992 calculated air concentration of 0.035 dpm/m3 was based on applying 

a resuspension factor of 1 × 10-6 m-1 (Abu-Eid et al. 2002) to the maximum removable surface 

contamination of 350 dpm alpha/100 cm2

. Using these two air concentrations, a source term 

depletion rate of 1.15 × 10-4 d-1 was calculated. The ingestion intake rates were calculated using the 

method described in Section 3.0. The estimated daily inhalation and ingestion intake rates to residual 

radioactivity from AEC operations at the site (Table 5-1), were calculated by assuming that workers 

were exposed for 2,000 hr/yr.

To reconstruct external exposure to residual radioactivity, the maximum reported exposure rate of 

0.014 mR/hr (Adams and Payne 1992a) was back-extrapolated using the source term depletion rate 

calculated from the internal data. The nonpenetrating exposure rate was determined by assuming 
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 23 of 32

Table 5-1. Annual internal and external exposure to residual radioactivity.

Year

Inhalationa

(pCi U/d)

Ingestionb

(pCi U/d)

Penetratingc

(rem)

Nonpenetratingc

(rem)

1950d 0.627 0.029 0.157 0.784

1951 0.627 0.029 0.157 0.784

1952 0.601 0.028 0.150 0.751

1953 0.576 0.027 0.144 0.720

1954 0.552 0.026 0.138 0.691

1955 0.530 0.024 0.132 0.662

1956 0.508 0.023 0.127 0.635

1957 0.487 0.023 0.122 0.609

1958 0.467 0.022 0.117 0.584

1959 0.448 0.021 0.112 0.560

1960 0.429 0.020 0.107 0.537

1961 0.412 0.019 0.103 0.515

1962 0.395 0.018 0.099 0.494

1963 0.379 0.018 0.095 0.473

1964 0.363 0.017 0.091 0.454

1965 0.348 0.016 0.087 0.435

1966 0.334 0.015 0.083 0.417

1967 0.320 0.015 0.080 0.400

1968 0.307 0.014 0.077 0.384

1969 0.294 0.014 0.074 0.368

1970 0.282 0.013 0.071 0.353

1971 0.270 0.013 0.068 0.338

1972 0.259 0.012 0.065 0.324

1973 0.249 0.012 0.062 0.311

1974 0.238 0.011 0.060 0.298

1975 0.229 0.011 0.057 0.286

1976 0.219 0.010 0.055 0.274

1977 0.210 0.010 0.053 0.263

1978 0.202 0.009 0.050 0.252

1979 0.193 0.009 0.048 0.242

1980 0.185 0.009 0.046 0.232

1981 0.178 0.008 0.044 0.222

1982 0.170 0.008 0.043 0.213

1983 0.163 0.008 0.041 0.204

1984 0.157 0.007 0.039 0.196

1985 0.150 0.007 0.038 0.188

1986 0.144 0.007 0.036 0.180

1987 0.138 0.006 0.035 0.173

1988e 0.132 0.006 0.033 0.166

1989 0.127 0.006 0.032 0.159

1990 0.122 0.006 0.030 0.152

1991 0.117 0.005 0.029 0.146

1992 0.112 0.005 0.028 0.140

1993e 0.107 0.005 0.027 0.134

1994e 0.103 0.005 0.026 0.129

1995 0.099 0.005 0.025 0.123

a. Absorption Types M and S are possible. 

b. Choose same f1-value as used for inhalation per NIOSH (2004).

c. External doses should be assessed using Exposure (R) DCFs.

d. The operational period ends on 02/28/1950. Therefore, the residual period is extended from 1951 back 

to 03/01/1950.

e. See text below for additional exposure scenarios from clean-up activities.
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 24 of 32

that the ratio of nonpenetrating to penetrating exposure rates for submersion and contamination 

external exposures during the operational exposure period (5 to 1), provided a reasonable estimate of 

the ratio of nonpenetrating to penetrating exposure rate during the residual exposure period. The 

estimated annual penetrating and nonpenetrating external exposures to residual radioactivity from

AEC operations at the site (Table 5-1) were calculated by assuming that workers were exposed for 

2,000 hr/yr. 

In addition to general exposure from residual contamination, two cleanup efforts (in 1988 and 1993 to 

1994) were performed that could result in additional exposure. The 1988 effort was limited in 

Building 3 and occurred in November and December of 1988. Vacuums were fitted with highefficiency particular air (HEPA) filters to clean the floors and walls. Contaminated bricks and soil were 

removed as necessary. In addition, respiratory protection equipment was used to reduce the 

likelihood of inhaling contaminated particulates. Further, workers were required to wear lapel air 

monitors that were analyzed every 24 hours. The lapel air data was not available for this analysis. 

The prejob estimates for the 1988 cleanup efforts were about 0.0034 pCi/m3 for about 150 hours (ANL

1988). There is air data from the cleanup in 1993 and 1994 at Aliquippa Forge (TMA/Eberline 1993). 

The air concentration for vacuuming tasks was 1 to 2 orders of magnitude greater than the other tasks 

and general area numbers. Therefore, for the 1988 cleanup effort, a bounding assumption of 100 

times the estimated residual air concentration was used to estimate internal exposure. Multiplying the 

1988 intake rate by 100 resulted in cleanup intake rates of 13.245 pCi/d for inhalation and 0.613 pCi/d 

for ingestion. These intake rates apply only to November and December of 1988.

The decontamination techniques in 1993 and 1994 were much more aggressive than in 1988. In 

addition to HEPA vacuuming, which was the main method in 1988, mechanical shot blasting, concrete 

saws, and jack-hammering were employed (DOE 1996b). The maximum air concentration during the 

1993 to 1994 cleanup was 1.1 × 10-10 μCi/mL. This air concentration was for a 3.75-hour job, 

recorded on a lapel air sampler, and without accounting for the worker’s use of respiratory protection

doing decontamination work on the west furnace (DOE 1996b). Using this air concentration as a 

bounding estimate of the decontamination and decommissioning work, the calculated intake rates 

were 724 pCi/d for inhalation and 33.5 pCi/d for ingestion. These intake rates apply only from June of 

1993 to September of 1994. 

6.0 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 25 of 32

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Edition, Academic Press, San Diego, California.

Harbert, R. R., 1989, “Bechtel Job No. 14501, FUSRAP Project, DOE Contract No. DE-AC05-

81OR20722, Post-Remedial Action Summary of the Aliquippa Forge 1988, Decontamination of 

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1948,” HW-10987, letter to T. W. Hauff (U.S. Atomic Energy Commission), General Electric 

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Commission, New York Operations Office, New York, New York, May 10. [SRDB Ref ID: 

6291, p. 93]

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OCAS-TIB-009, Rev. 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, April 

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for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, DCAS-IG-003, Rev. 1, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, October 5.

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Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 30 of 32

GLOSSARY

absorption type

Categories for materials according to their rate of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorbtion into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively in the 

respiratory tract (slow solubilization). Also called solubility type.

activity median aerodynamic diameter (AMAD)

Diameter of a unit density sphere with the same terminal settling velocity in air as that of the 

aerosol particle whose activity is the median for the entire aerosol.

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2.

anterior–posterior (AP)

Physical orientation of the body relative to a penetrating directional radiation such that the 

radiation passes through the body from the front to the back.

atomic weapons employer (AWE) [42 U.S.C. § 7384l(5)]

Entity other than the United States, that—(A) processed or produced, for use by the United 

States, material that emitted radiation and was used in the production of an atomic weapon, 

excluding uranium mining and milling, and (B) is designated by the Secretary of Energy as an 

atomic weapons employer for purposes of the [Energy Employees Occupational Illness] 

compensation program.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

biological half-life

Time required for the body or an organ to eliminate, by biological processes, one half of the 

amount of an ingested substance.

bremsstrahlung

Electromagnetic radiation released as a result of inelastic scattering of a moving charged 

particle within the nucleus of an atom. X-rays produced in a typical medical X-ray tube 

frequently originate from inelastic scattering of accelerated electrons in the anode material.

breathing rate

Amount of air a person breathes in a specified time. In relation to health physics for workers, 

rates typically vary from light (1.2 cubic meters per hour) to heavy (1.7 cubic meters per hour) 

as defined by the International Commission on Radiological Protection.

contamination

Radioactive material in an undesired location including air, soil, buildings, animals, and 

persons.
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 31 of 32

counts per minute (cpm)

Unit of measurement on a radiation meter in which a count indicates a detected ionization 

event.

disintegrations per minute (dpm, d/m)

Measure of radioactivity equal to the number of nuclear disintegrations in a mass per minute; 

1 dpm equals 1/60 becquerel.

dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of roentgens, rads, reps, or grays.

electron-volt (eV)

Unit equal to the energy of one electron moving through a potential difference of 1 volt 

(1.602 × 10-19 joules). The common units in nuclear physics and radiology are kiloelectronvolts (thousands) and megaelectron-volts (millions).

exposure

(1) In general, the act of being exposed to ionizing radiation. See acute exposure and chronic 

exposure. (2) Measure of the ionization produced by X- and gamma-ray photons in air in units 

of roentgens.

favorable to claimants

In relation to dose reconstruction for probability of causation analysis, having the property of 

ensuring that there is no underestimation of potential dose, which often means the assumption 

of a value that indicates a higher dose than is likely to have actually occurred in the absence of 

more accurate information. See probability of causation.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus.

high-efficiency particulate air (HEPA) filter

Dense filter that removes contaminants from air flows before return to the working environment 

or discharge to the outside air (exhaust).

Interactive RadioEpidemiological Program (IREP)

Computer program that uses a person’s calculated annual organ doses and other information 

(e.g., gender, age at diagnosis, and age at exposure) to calculate the probability of causation 

of a specific cancer for a given pattern and level of radiation exposure.

probability of causation (POC)

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Act, the percent likelihood, at the 99th percentile, that a worker incurred a 

particular cancer from occupational exposure to radiation.

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joules per kilogram) of absorbing tissue. The rad has 
Document No. ORAUT-TKBS-0021 Revision No. 01 Effective Date: 04/26/2012 Page 32 of 32

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose.

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei.

radon (Rn)

Radioactive gaseous element with atomic number 86. Radon is a decay product (progeny) of 

other radioactive elements such as thorium and radium.

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

rep

Historical quantity of radiation (usually other than X-ray or gamma radiation) originally defined 

as 83 ergs absorbed per gram in the body and redefined in the 1940s or early 1950s as the 

amount that would liberate the same amount of energy (93 ergs per gram) as 1 roentgen of Xor gamma rays. Replaced by the gray in the International System of Units; 1 rep is 

approximately equal to 8.38 milligray. The word derives from roentgen equivalent physical.

roentgen (R, sometimes r)

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0°C and standard atmospheric pressure. An 

exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher energy photons (generally greater than 100 kiloelectron-volts).

U.S. Atomic Energy Commission (AEC)

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 

Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the U.S. Energy Research and 

Development Administration in 1979.

uranium (U)

Heavy, metallic, and radioactive element with atomic number 92. Most natural uranium as 

found in ores is 238U with trace levels of other isotopes. Uranium-235 (0.7% of natural 

uranium) is fissile by itself and used in nuclear weapons as well as reactors. Uranium-238 

(99.3% of natural uranium) is fissionable by fast neutrons and used in nuclear reactors. 

Natural uranium contains a minute amount of 234U.

working level

Unit of concentration in air of the short-lived decay products of 222Rn (218Po, 214Pb, 214Bi, and 

214Po) and 220Rn (216Po, 212Pb, 212Bi, 212Po) defined as any combination of the short-lived 

radioactive progeny of radon or thoron in 1 liter of air, without regard to the degree of 

equilibrium, that results in the ultimate emission of 130,000 MeV of alpha energy; 1 WL equals 

2.083 × 10-5 joules per cubic meter.